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Abstract: Depth profiles of deuterium trapped in graphite tiles not exposed to hydrogen discharges just before air ventilation of JT-60U were determined based on the D(3He,p)4He resonant nuclear reaction analysis. The samples were taken from the W-shaped divertor. The maximum depth of analysis is about 1.4×1024 carbon atoms/m2, corresponding to a linear depth of 16μm for the density of 1.7×103 kg/m3. The highest concentration (19at.%) was found at the surface of the inner divertor area, which was significantly higher than that exposed to H discharges at the same location. However in the deeper region, the deuterium concentrations without H discharges were lower than those with H discharges in 3 of 4 samples. These results indicate that hydrogen discharges can remove deuterium trapped only in the shallow region. [Copyright &y& Elsevier]

Abstract: The possibility of tritium self-sufficiency in a low aspect ratio (low-A) tokamak reactor with the outboard blanket only has been studied to simplify the inboard structure located in a high magnetic field. The tritium breeding ratio (TBR) of the outboard blanket increases by applying a neutron multiplier such as lead and beryllium to the inboard reflector. The effect in local TBR of a 40cm thick inboard reflector of lead is about 40% that of the inboard breeding blanket. The local TBR of a reactor with both the lead inboard reflector and outboard blanket using a Li2O breeder and a beryllium multiplier can be larger than 1.35 for aspect ratios less than 2.9. The local TBR of 1.35 satisfies the net TBR requirement of 1.05 where there is 78% coverage of the breeding zone over the total plasma-facing surface area. The result indicates there is a design solution achieving tritium self-sufficiency even in a low-A tokamak reactor with the outboard breeding blanket only, using an appropriate inboard reflector. Be12Ti and Li2TiO3 are recommended from the viewpoint of safety. In the combination, the aspect ratio is required to be less than 1.35 to achieve local TBR of 1.35. [Copyright &y& Elsevier]

Abstract: Deuterium concentrations and depth profiles in plasma-facing graphite tiles used in the divertor of JAERI Tokamak-60 Upgrade (JT-60U) were investigated by nuclear reaction analysis (NRA). The highest deuterium concentration of D/12C of 0.053 was found in the outer dome wing tile, where the deuterium accumulated probably through the deuterium–carbon co-deposition. In the outer and inner divertor target tiles, the D/12C data were lower than 0.006. Additionally, the maximum (H+D)/12C in the dome top tile was estimated to be 0.023 from the results of NRA and secondary ion mass spectroscopy (SIMS). Orbit following Monte-Carlo (OFMC) simulation showed energetic deuterons caused by neutral beam injections (NBI) were implanted into the dome region with high heat flux. Furthermore, the surface temperature and conditions such as deposition and erosion significantly influenced the accumulation process of deuterium. The deuterium depth profile, scanning electron microscope (SEM) observation and OFMC simulation indicated the deuterium was considered to accumulate through three processes: the deuterium–carbon co-deposition, the implantation of energetic deuterons and the deuterium diffusion into the bulk. [Copyright &y& Elsevier]

It is important to measure retentions and depth profiles of deuterium and tritium on plasma facing materials to understand plasma-wall interactions. The nuclear reaction analysis (NRA) has been established using a deuteron beam with an adjustable energy. By means of the NRA, depth profiles of deuterium and tritium on the surface of JT-60 divertor graphite tiles irradiated with DD burning were obtained. The NRA method showed that the average concentrations of the deuterium retained in the surface of two different were 1.3 × 1016 and 0.3 × 1016 cm-2. α-particles from the T(d, α)n reaction were measured from a dome unit tile of JT-60U W-shaped divertor and the measured activity of tritium was about 18 kBq/cm2. The peak energy of the α spectrum obtained with the NRA was about 1.2 MeV. This result implies that energetic tritons were implanted in the divertor graphite tile. [Copyright &y& Elsevier]

Abstract: Dust generated by plasma–wall interaction is a potential source of tritium retention in a fusion reactor. Evaluation of the dust accumulation in the entire vacuum vessel is required to estimate the total amount of tritium retention, but it was particularly difficult to measure for plasma-unexposed areas behind the PFC structures, i.e. “shadow areas”. Dust samples were collected at 3, 5 and 2–4 different toroidal locations on the first wall, divertor surface and the exhaust route under the divertor in JT-60U, respectively. On the tile surface, large mass area density was found at the inner divertor and baffle, in particular, upper tiles compared to the lower target tile where the thick deposition layers were produced. Mass area density was significantly increased at the shadow areas, i.e. under the divertor structure such as the divertor and baffle tiles and the divertor dome. It was found that the poloidal distribution is relatively symmetrical in the toroidal direction within a factor of three. In comparison with the previous collection just before major change of the plasma operations, dust accumulation was increased both at the exposed and shadow areas due to change in the operating conditions. [Copyright &y& Elsevier]

Abstract: The ventilation systems design for the ITER nuclear buildings ensures radioactive contamination is confined so that workers, the public and the environment are protected. Nuclear buildings are divided into confinement sectors which connect to the Heating, Ventilation and Air Conditioning (HVAC) system and detritiation system (DS). The Tokamak Complex DS provides centralized air purification for the building confinement sectors. A distributed arrangement of ventilation piping provides networks necessary for two key functions, these being Vent Detritiation (VD), to maintain sub-atmospheric pressure, and Air Detritiation (AD) to collect tritium released into the confinement sector. For the VD function, air extracted from the particular confinement sector is directed to the DS for processing prior to exhaust to the environment. This paper presents the configuration of the DS of the Tokamak Complex and addresses details of the design of the distributed piping network. Dynamic flow and pressure drop modelling has been applied to support the development of the system configuration and provide data for sizing the system and selecting components. Further design development is discussed in view of the safety requirements for operation of the system during design basis events such as earthquake or fire. [ABSTRACT FROM AUTHOR]

Abstract: The ITER Nuclear Buildings include the Tokamak, Tritium and Diagnostic Buildings (Tokamak Complex) and the Hot Cell and Low Level Radioactive Waste Buildings (Hot Cell Complex). The Tritium Confinement strategy of the Nuclear Buildings comprises key features of the atmosphere and vent detritiation systems and the heating, ventilation and air conditioning systems. The designs developed during the ITER EDA (engineering design activities) for these systems need to be adapted to the specific conditions of the Cadarache site and modified to conform with the regulatory requirements applicable to installations nucléaires de base (INB) – basic nuclear installations – in France. The highest priority for such adaptation has been identified as the Tritium Confinement of the Tokamak Complex and the progress in development of a robust, coherent design concept compliant with French practice is described in the paper. [Copyright &y& Elsevier]

This paper describes the design of the ITER tritium plant subsystems, layout in the tritium building and the construction plan. The tritium plant comprises tokamak fuel cycle processing systems, as well as tritium confinement and detritation systems. The plant processes tritiated gases received from the tokamak and other sources to produce the D, T gas streams for fuelling, and detritiates various waste streams including tritiated water before discharge to the environment. The plant has been designed to meet not only all anticipated plasma operation scenarios in the DD and DT phases with a wide range of burn pulse durations from short pulse (450 s) and long pulse (3000 s), but also safety requirements (minimization of equipment tritium inventory and environmental tritium release from different accidental events in tokamak and tritium processing subsystems, and reduction of workers’ tritium exposure, etc). [Copyright &y& Elsevier]

Abstract: JT-60SA is a fully superconducting coil tokamak to be built under the framework of the EU-JA Broader Approach Agreement, and it aims to contribute to the complement of the ITER experiments and to the DEMO reactor design by the study of steady-state high-beta plasma experiments. The conceptual design of the JT-60SA tokamak and the peripheral systems has been carried out in close collaboration of EU and Japan from middle of 2006, and the JT-60SA Conceptual Design Report which involves an analysis of plasma physics and engineering design is summarized in May 2007. This paper intends to present an overall view of the JT-60SA project, and the latest design status of the key engineering issues such as superconducting magnets, divertor target, plasma heating devices, and magnet power supplies. [Copyright &y& Elsevier]

Nonlinear simulations based on the magnetohydrodynamic model have been executed to reveal the dynamics of the ballooning mode in the spherical tokamak plasma. The simulation results have reproduced the characteristic features of the edge-localized mode crash phase, where the filamentary structures are formed along the magnetic field in the edge region, and separated from the core plasma. Moreover, the finite Larmor radius effect is addressed. [ABSTRACT FROM AUTHOR]

Low-Z impurity (7Be) on the JT-60U divertor tiles was analyzed to study the impurity behavior in the divertor region. The amount of the 7Be increased approximately one hundred times after B4C-tile installation in the outer divertor. The 7Be was probably produced by 10B(p,α)7Be nuclear reaction on the divertor tiles in the hydrogen experiment with ion cyclotron range of frequency heating. The 7Be was distributed asymmetrically in the poloidal and the toroidal direction. The highest 7Be concentration was found at the inner divertor whose boron content (B/(B+C) ∼ 20%) was lower than the B4C tiles (B/(B+C) ∼ 80%) of the outer divertor. This result may imply impurity transport from the outer divertor to the inner divertor. [Copyright &y& Elsevier]